1. Briesmeister, J.F., 1993. MCNP: A General Monte Carlo N-Particle Transport Code. Los Alamos National Laboratory.
2. Chadwick, M.B., Young, P.G., Fu, C.Y., June 1996. ENDF/B-VII.1 MAT 125.
3. Chadwick, M.B., Young, P.G., Fu, C.Y., September 1996. ENDF/B-VII.1 MAT 2631.
4. ENDF/B-VII.1 nuclear data for science and technology: cross sections, covariances, fission product yields and decay data;Chadwick;Nucl. Data Sheets,2011
5. Cullen, D.E., May 2014. How Accurate Are Our Processed ENDF Cross Sections?. Technical report. IAEA Nuclear Data Section.